Stability of ferritic steel to higher doses: Survey of reactor pressure vessel steel data and comparison with candidate materials for future nuclear systems

doi:10.1016/j.ijpvp.2014.06.001

International Journal of Pressure Vessels and Piping

Volume 122, October 2014, Pages 1-5

https://doi.org/10.1016/j.ijpvp.2014.06.001 Get rights and content

Highlights

•
Common embrittlement trend between RPV and advanced steels till intermediate doses.
•
For doses >1.5 dpa, damage rate saturation tendency is observed for RPV steels.
•
RPV steels might be conveniently utilised also outside their foreseen dose range.

Abstract

This paper is illustrating the potential of the well-known low alloyed clean steels, extensively used for the current light water Reactor Pressure Vessels (RPV) steels, for a likely use as a structural material also for the new generation nuclear systems. This option would provide, especially for large components, affordable, easily accessible and a technically more convenient solution in terms of manufacturing and joining techniques.

A comprehensive comparison between several sets of surveillance and research data available for a number of RPV clean steels for doses up to 1.5 dpa, and up to 12 dpa for 9%Cr steels, is carried out in order to evaluate radiation stability of the currently used RPV clean steels even at higher doses. Based on the numerous data available, positive preliminary conclusions are drawn regarding the eventual use of clean RPV steels for the massive structural components of the new reactor systems.

Section snippets

Background

Ferritic steels have been considered as candidate structural materials for fusion power plants since the late 1970s. The reason being is that the data obtained from fast reactor irradiation shows that ferritic steels are more swelling resistant than austenitic stainless steels. Moreover, their higher thermal conductivity and lower thermal expansion coefficients lead to improved resistance to thermal stresses characteristic for a fusion power plant operating in a pulsed mode.

Mainly high Cr

Utilised data and assumptions for their comparison

Critical sets of surveillance PWR data are used for this study. Representative clean ferritic western type PWR RPV steels [22] are selected. The upper fluence is approximately corresponding to 80–90 mdpa (milli-dpa). Additionally, surveillance data at ∼270 °C taken from Russian type pressurised water reactor WWER-440 type are also considered in the study. An important remark should be made that only ‘clean’ steels with sufficiently low Cu and P content are utilised. The fluence range for

Additional dataset for comparison

Very comprehensive research has been done on using different Cr content steels for fusion applications. Mainly modified 2.25%Cr and higher Cr steels (up to 12%Cr) have been considered as potential candidates [1], [2], [3], [4], [5], [6], [7], [8], [9]. High dose embrittlement data obtained at 365 °C irradiation in Fast Flux Test Facility (FFTF), owned by the U.S. Department of Energy, are available in the literature [9]. The data and their trend are used for comparison in relation to the

Data comparison and modelling implications

Since relatively clean steels were considered for the study, only basic matrix damage (material hardening occurring during irradiation) was taken into account in the embrittlement process. Embrittlement data on low Cr steels PWR, BWR and WWER 440 steels, and high Cr (up to 9%Cr) Eurofer 97 and 9Cr2WVTa steels, represented as an increase in the tensile yield stress as a function of the dose are shown in Fig. 1. For better representation a logarithmic scale was used. Based on a physically sound

Chromium effect on irradiation stability

Embrittlement data, expressed again in tensile yield stress increase as a function of the dose, at higher doses (up to 28 dpa) and higher irradiation temperature of 365 °C for different Cr content steels are analysed and compared in Fig. 2 to the established embrittlement trend shown in Fig. 1. Evidence for chromium stabilisation of irradiation damage in RPV weld metals is investigated and shown in Ref. [36]. The measured DBTT shift for high-Cr welds lie well below the value measured for low Cr

Conclusions

Manufacturing of extremely large and heavy pressure vessels using expensive non-conventional material and complicated joining procedures is not sustainable whereas the utilisation of the best conventional well proven materials, like LWR RPV steels may offer significant advantages. Using conventional low Cr clean RPV steel might be a smart, economical and feasible solution for pressure vessel and other massive components of the future nuclear reactors.

Surveillance data from WWER-440 materials up

References (36)

R.L. Klueh et al.
Impact behavior of 9-Cr and 12-Cr ferritic steels after low-temperature irradiation
J Nucl Mater
(1988)
R.L. Klueh et al.
Impact behavior of 9Cr-1MoVNb and 12Cr-1MoVW steels
J Nucl Mater
(1991)
R.L. Klueh
Chromium–molybdenum steels for fusion reactor first Walls – a review
Nucl Eng Des
(1982)
R.L. Klueh et al.
Development of low-Cr, Cr-W steels for fusion
J Nucl Mater
(1995)
R.L. Klueh et al.
Effect of heat treatment and irradiation temperature on impact properties of Cr-W-V ferritic steels
J Nucl Mater
(1999)
R.L. Klueh et al.
Impact behavior of reduced-activation steels irradiated to 24 dpa
J Nucl Mater
(1996)
R.L. Klueh
Creep-rupture behaviour of a bainitic 2.25Cr1Mo steel
Int J Press Vessels Pip
(1980)
L. Debarberis et al.
Role of nickel in a semi-mechanistic analytical model for radiation embrittlement of model alloys
J Nucl Mater
(2005)
R.E. Stoller et al.
Primary damage formation in BCC iron
J Nucl Mater
(1997)
A.J. Magielsen et al.
Irradiation of structural materials in contact with lead bismuth eutectic in the high flux reactor
J Nucl Mater
(2011)

A. Kryukov et al.

Prediction of irradiation embrittlement of vanadium alloyed low nickel steel for future reactors

J Nucl Mater

(2011)

L. Debarberis et al.

Effect of irradiation temperature in PWR RPV materials and its inclusion in semi-mechanistic model

Scr Mater

(2005)

G.M. Kalinin et al.

Specification of stress limits for irradiated 316L(N)-IG steel in ITER structural design criteria

J Nucl Mater

(2004)

K.L. Murty et al.

Structural materials for Gen-IV nuclear reactors: challenges and opportunities

J Nucl Mater

(2008)

L. Debarberis et al.

Evidence for chromium stabilisation of irradiation damage in RPV weld metals

Int J Press Vessels Pip

(2006)

W.L. Hu et al.

R.L. Klueh et al.

Tensile behavior of three commercial ferritic steels after low-temperature irradiation, Proceedings of the topical conference on ferritic alloys for use in nuclear energy technologies

(1984)

R.L. Klueh et al.

A comparison of low-chromium and high-chromium reduced-activation steels for fusion applications

Cited by (13)

Computational Kinetics: Application to Nuclear Materials
2020, Comprehensive Nuclear Materials: Second Edition
Allotropic transformations involving interface diffusion and precipitation are driving many properties controlling the lifespan of nuclear power plant parts. Predictive models at the scale of the microstructure (as part of a more general multi-scale approach) are useful to better understand and optimize the microstructure of nuclear materials. After a brief description of thermo-kinetics basic concepts (diffusion driven interface motion and classical nucleation and growth theories), recent applications of phase transformation kinetics to nuclear materials are presented: from allotropic transformations in steels and Zr alloys to the precipitation of oxides, copper, cementite in steels and β_Nb in Zr.
Microstructural characterization of the synergic effects of dynamic strain ageing and hydrogen on fracture behaviour of low-alloy RPV steels in high-temperature water environments
2020, Materials Characterization
Citation Excerpt :
Commercial RPVs are made of low-alloy steels (LAS) with typical upper bainite microstructure, such as MnMoNi and NiMoCr steel grades [1]. Irradiation embrittlement [2–5], as well as the high-temperature water (HTW) environment and the absorption of hydrogen from the reactor coolant and corrosion reactions can lead to the reduction of the fracture resistance of RPV. The degradation effects of HTW environment and hydrogen on the fracture properties of RPV steels can be concomitant with the effects of irradiation embrittlement [6], dynamic strain ageing (DSA) [7,8] or environmentally-assisted cracking (EAC) [9–11].
Tensile tests in high-temperature air with pre-charged hydrogen and elastic plastic fracture mechanics tests in hydrogenated high-temperature water (HTW) at 250 and 288 °C on low-alloy reactor pressure vessel (RPV) steels revealed a clear but moderate reduction of ductility and fracture resistance, respectively. The observed behaviour is a consequence of synergistic effects between hydrogen embrittlement (HE) and the dynamic strain ageing (DSA), in which the HE was amplified by a high DSA susceptibility. The deformation microstructures in the vicinity of the crack tips in air and HTW of two RPV steels with high DSA susceptibility were characterized in detail. These investigations support the idea that the environmental degradation of fracture resistance in hydrogenated HTW was mainly due to the plasticity localization by the interaction between DSA and hydrogen in RPV steels. Synergistic effects of DSA and hydrogen lead to heterogeneous distribution of dislocations and formation of dislocation cells inside bainitic laths.
On the evolution of mechanical properties and microstructure of ferritic-bainitic (FB) 2.25Cr-1Mo (Grade 22) steel during high-temperature creep
2020, Materialia
Citation Excerpt :
In a typical application, Grade 22 steel is used for superheater tubing as well as a filler metal for joining steam piping. Due to its excellent creep-rupture strength [4,5] and radiation resistance [6,7], it is also considered as a candidate material for application in the next generation (Generation IV) of nuclear reactors [8] as a pressure vessel material [9]. Because of its use at high temperatures under loading, it is of technological importance to monitor creep damage of components made of Grade 22 steel in order to determine the remaining creep life of a component under operation conditions.
The development of creep damage in ferritic-bainitic 2.25Cr-1Mo (Grade 22) steel was studied by following the evolution of mechanical properties and microstructure by characterisation of a series of interrupted creep specimens (605 °C/80 MPa). The change in mechanical properties of the partially crept specimens was evaluated via hardness measurements, while the microstructural changes were characterized by a range of microscopy techniques. The nano-hardness measurements combined with a statistical analysis allowed for analysis of the strength of the present constitutive phases as a function of creep time (t_c), and imparted creep strain (ɛ_c). In addition, a machine learning methodology was used to identify ferrite and bainite regions in Electron Back-Scatter Diffraction (EBSD) orientation maps, which then allowed for a study of the evolution of dislocation densities (Geometrically-Necessary Dislocations, GNDs) in each phase separately. It is shown that the amount of GNDs in the bainite is reduced significantly during the high-temperature creep, while there is a moderate increase in the amount of GNDs in ferrite grains. These trends are consistent with the observed changes in the strength (hardness) of each phase – bainite is significantly softening, while ferrite slightly hardening. The reduction of the amount of dislocations in the bainite and its consequent softening is caused by the progressive decomposition of the needle-like bainite present in the initial microstructure into polygonal fine-grained ferrite-like bainite. Transmission electron microscopy (TEM) confirmed the changes in dislocation density observations and was also used to identify changes in carbides during the high-temperature creep.
Environmental degradation of fracture resistance in high-temperature water environments of low-alloy reactor pressure vessel steels with high sulphur or phosphorus contents
2019, Corrosion Science
Citation Excerpt :
The body of the RPV is made of fine-grained low alloy steel (LAS) with granular upper bainite microstructure, such as MnMoNi or NiMoCr steel series [2,3]. The fracture resistance of RPV is mainly degraded by irradiation embrittlement [4–6] due to neutrons and thermal ageing [7], and potentially by the high-temperature water (HTW) environment and the absorption of hydrogen. The hydrogen absorption from the reactor coolant is due to radiolysis, intentional additions or corrosion reactions.
The fracture behaviour of low-alloy reactor pressure vessel steels with various sulphur and phosphorus contents, different environmentally-assisted cracking (EAC) and temper embrittlement (TE) susceptibilities was evaluated by elastic plastic fracture mechanics tests in air and various simulated light water reactors environments. A moderate but clear reduction of fracture initiation resistance occurred in a) high-sulphur steel with high EAC susceptibility in oxygenated high-temperature water with aggressive occluded crevice environment and preceding EAC growth, and in b) high-phosphorus steel with high TE susceptibility, where the reduction of fracture resistance was most pronounced in hydrogenated high-temperature water.
Effect of dynamic strain ageing on environmental degradation of fracture resistance of low-alloy RPV steels in high-temperature water environments
2019, Corrosion Science
Citation Excerpt :
Commercial RPVs are made of fine-grained low-alloy steels with granular upper bainite microstructure, such as MnMoNi and NiMoCr steel grades [1]. The fracture resistance of RPV can be reduced by irradiation embrittlement due to neutrons and thermal ageing [2–4], as well as potentially by the high-temperature water (HTW) environment and the absorption of hydrogen from the reactor coolant and corrosion reactions. The fracture behaviour of RPV steels in HTW is expected to be affected by material (strength, chemical composition, DSA susceptibility, microstructure, etc.), environmental (temperature, water chemistry, etc.) and mechanical loading parameters (strain rate, constraints, etc.).
Elastic plastic fracture mechanics tests on different low-alloy reactor pressure vessel (RPV) steels in air and various simulated light water reactor environments revealed a clear, but moderate reduction of the fracture initiation resistance of RPV steels with high susceptibility to dynamic strain ageing (DSA) in high-temperature water (HTW). In HTW, the reduction of fracture initiation resistance increased with decreasing loading rate and was most pronounced in hydrogenated HTW. The stronger reduction of fracture resistance in hydrogenated HTW was mainly due to the localization of plastic deformation by interaction between DSA and hydrogen.
Motivation for utilizing new high-performance advanced materials in nuclear energy systems
2016, Current Opinion in Solid State and Materials Science
Citation Excerpt :
This reduction in thickness, and therefore ingot mass, would greatly reduce fabrication costs while enabling access to numerous smaller capacity vendors (existing in the US and elsewhere). Furthermore, safety and reliability may be enhanced for this critical component given the anticipated improved irradiation resistance in these alloys [42–44]. For example, the unirradiated fracture toughness for 3Cr-3WV steel is similar or superior to conventional 1Cr RPV steels [40], and the irradiation-induced shift in ductile to brittle transition temperature and reduction in upper shelf toughness is expected to be markedly less for the 3Cr-3WV steels due to reduced radiation hardening [10,11], although additional experimental data are needed to confirm this expectation.
Despite the very demanding operational environment in nuclear reactors, there have been relatively few advanced high-performance materials introduced into fission reactors during the past 50 years. Some of the regulatory and operational barriers to the introduction of high performance materials are briefly discussed, and several examples of potential improvement in current and planned fission reactor systems that could be enabled by advanced structural materials for in-core applications are outlined. Enhanced non-proprietary public-private research and development on advanced structural materials could yield numerous performance, economic, environmental and safety benefits for current boiling water and pressurized water reactors as well as future Generation IV reactor systems such as sodium cooled fast reactors or very high temperature gas cooled reactors.

View all citing articles on Scopus

View full text

Stability of ferritic steel to higher doses: Survey of reactor pressure vessel steel data and comparison with candidate materials for future nuclear systems

Highlights

Abstract

Section snippets

Background

Utilised data and assumptions for their comparison

Additional dataset for comparison

Data comparison and modelling implications

Chromium effect on irradiation stability

Conclusions

J Nucl Mater

J Nucl Mater

Nucl Eng Des

J Nucl Mater

J Nucl Mater

J Nucl Mater

Int J Press Vessels Pip

J Nucl Mater

J Nucl Mater

J Nucl Mater

J Nucl Mater

Scr Mater

J Nucl Mater

J Nucl Mater

Int J Press Vessels Pip

Tensile behavior of three commercial ferritic steels after low-temperature irradiation, Proceedings of the topical conference on ferritic alloys for use in nuclear energy technologies

A comparison of low-chromium and high-chromium reduced-activation steels for fusion applications